Long-term development of nuclear energetics is associated with production of fast power reactors that can allow solving crucial problems of effective and safe usage of nuclear fuel upon closure of the nuclear fuel cycle and providing environmental safety. Ongoing efforts include the development of new generation lead-cooled fast reactors having nitride uranium-plutonium fuel. Such reactors show fundamental advantages over the thermal reactors and sodium-cooled fast reactors and can create the basis for development of power plants characterised in high level of reliability, safety and substantially unlimited fuel supply. (The next generation of fast reactors/E. O. Adamov, V. V. Orlov, A. I. Filin, V. N. Leonov, A. G. Sila-Novitsky, V. S. Smirnov, V. S. Tsikunov//Nuclear Engineering and Design.—1997.—Vol. 173, No 1-3.—P. 143-150).
There are engineering and design elaborations of a pool reactor system having master equipment of integral-loop-type based on liquid-lead-cooled fast reactor, described in the experimental test project of the reactor system BREST-OD-300 (V. N. Leonov, A. A. Pikalov, A. G. Sila-Novitsky et al. Design and layout solutions of major components and equipments of the reactor BREST-OD-300//Problems of Atomic Science and Technology. Series: Nuclear Physics Investigations, N 4, Moscow, State Unitary Enterprise Research and Development Institute of Power Engineering (NIKIET), 2004, p. 65-72).
The system includes a reinforced concrete cavity having an inner steel cladding, a reactor vessel unit with an upper cover, an active zone, an actuating mechanism system controlling the active zone reactivity, steam generators and main circulation pumps, a system of mass-transfer devices and filters for coolant purification, a refuelling system for active zone components, a process parameter monitoring system and other auxiliary systems. The vessel unit of the BREST-OD-300 reactor system comprises one central and four peripheral flat-bottomed tubular cavities which together with the upper cover define a primary circuit boundary of the reactor system where circulating therein coolant removes heat from the active zone, the volume of cover gas is generated and intrareactor devices and equipment are arranged. The active zone is arranged within the central cavity of the vessel unit, and steam generator units are arranged within four peripheral cavities connected to the central cavity by means of upper and lower pipes. Each steam generator has the form of a tubular heat exchanger for heating water (steam) of supercritical parameters which is immersed into the lead coolant flow circulating in a space between the pipes of the steam generator shell in the top-to-bottom direction. In the case of pipe leakage and steam flowing into the lead coolant circulation circuit, the steam generator switching off is provided by means of cutting off the feed-water and main-steam supply conduits in a secondary circuit. Lead coolant circulation within the reactor BREST-OD-300 is provided by means of pumping said coolant with circulation pumps out of a steam generator cavity to the level of a pressure chamber of the reactor, wherefrom said coolant descends to an active zone inlet chamber, then rises and is heated in the active zone in the result of contact with fuel elements of fuel assemblies and after that enters a “hot” coolant common chamber. Further, the coolant flows into the inlet chambers and the space between pipes of steam generators, is cooled down and is transferred to the circulation pump inlets; after which it is again supplied into the reactor pressure chamber.
The disclosed system structure is intended for providing a test reactor BREST-OD-300 and checking technical solutions that may be used in the development of new generation lead-cooled fast power reactors. Such reactors comprise master equipment of integral-loop-type which has large dimensions and high specific weight of used lead coolant per unit of output power. For instance, for the BREST-1200 reactor this factor is 1.4 m3/MW and more, which depends on unit power of the master equipment and layout arrangements.
As the closest solution was chosen a pool-type nuclear power system having master equipment of integral-loop-type based on a lead-cooled fast reactor (Ru 2247435). The system comprises a reactor arranged within a central tank, steam generators and circulation pumps arranged within peripheral tanks, and a coolant treatment system using gaseous mixtures for lead oxide recovery. The reactor, steam generators and circulation pumps are arranged under the liquid-metal-coolant free level. Steam generators of said system made in the form of a tubular heat exchanger having the pipes supplied with water (steam) and a space between the pipes where the lead coolant circulates in the top-to-bottom direction. Inside the reactor system, between the free liquid-metal-coolant level and an upper cover there is a common gaseous chamber communicated with a gas circulation and purification system.
A disadvantage of this technical solution is that the high inner pressure equipment (steam generators) is arranged within the peripheral tanks filled with molten lead. This increases the risk of accidents upon steam generator pipe integrity loss and ingression of water (steam) into coolant. Moreover, the integral-loop-type master equipment has high specific weight of used lead coolant per unit of reactor power which results in increase of reactor dimensions and capital costs for such reactor production.